SAS3D analysis of natural convection boiling behavior in the Sodium Boiling Test Facility
Conference
·
OSTI ID:5395827
The objective of the initial phase of testing in the Sodium Boiling Test (SBT) Facility, at the Oak Ridge National Laboratory, was to determine the maximum power that could be transferred by a simulated breeder reactor coolant subchannel when the coolant flow is driven by natural convection. In order to aid in the evaluation of the experimental data and to help understand the flow regimes present at the various power levels examined during this test program, a SAS3D computer model of the SBT Facility was developed.
- Research Organization:
- Oak Ridge National Lab., TN (USA); Argonne National Lab., IL (USA)
- DOE Contract Number:
- W-7405-ENG-26
- OSTI ID:
- 5395827
- Report Number(s):
- CONF-791103-92
- Country of Publication:
- United States
- Language:
- English
Similar Records
Natural convection boiling of sodium in a simulated FBR fuel assembly subchannel
SAS3A analysis of natural convection boiling behavior in the Sodium Boiling Test Facility
Sodium natural convection testing in the Thermal-Hydraulic Out-of-Reactor Safety (THORS) facility. [LMFBR]
Conference
·
Sun Dec 31 23:00:00 EST 1978
·
OSTI ID:5902714
SAS3A analysis of natural convection boiling behavior in the Sodium Boiling Test Facility
Conference
·
Sun Dec 31 23:00:00 EST 1978
·
OSTI ID:5902707
Sodium natural convection testing in the Thermal-Hydraulic Out-of-Reactor Safety (THORS) facility. [LMFBR]
Conference
·
Wed Dec 31 23:00:00 EST 1980
·
OSTI ID:6425099
Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
210500 -- Power Reactors
Breeding
22 GENERAL STUDIES OF NUCLEAR REACTORS
220600 -- Nuclear Reactor Technology-- Research
Test & Experimental Reactors
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
AFTER-HEAT
AFTER-HEAT REMOVAL
BREEDER REACTORS
COMPUTER CALCULATIONS
CONVECTION
ENERGY TRANSFER
EPITHERMAL REACTORS
FAST REACTORS
FBR TYPE REACTORS
FFTF REACTOR
FLUID MECHANICS
HEAT TRANSFER
HYDRAULICS
LIQUID METAL COOLED REACTORS
LMFBR TYPE REACTORS
LOSS OF FLOW
MECHANICS
MOCKUP
NATURAL CONVECTION
REACTOR ACCIDENTS
REACTOR SAFETY
REACTORS
REMOVAL
RESEARCH AND TEST REACTORS
RESEARCH REACTORS
SAFETY
SODIUM COOLED REACTORS
STRUCTURAL MODELS
TEST FACILITIES
TEST REACTORS
210500 -- Power Reactors
Breeding
22 GENERAL STUDIES OF NUCLEAR REACTORS
220600 -- Nuclear Reactor Technology-- Research
Test & Experimental Reactors
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
AFTER-HEAT
AFTER-HEAT REMOVAL
BREEDER REACTORS
COMPUTER CALCULATIONS
CONVECTION
ENERGY TRANSFER
EPITHERMAL REACTORS
FAST REACTORS
FBR TYPE REACTORS
FFTF REACTOR
FLUID MECHANICS
HEAT TRANSFER
HYDRAULICS
LIQUID METAL COOLED REACTORS
LMFBR TYPE REACTORS
LOSS OF FLOW
MECHANICS
MOCKUP
NATURAL CONVECTION
REACTOR ACCIDENTS
REACTOR SAFETY
REACTORS
REMOVAL
RESEARCH AND TEST REACTORS
RESEARCH REACTORS
SAFETY
SODIUM COOLED REACTORS
STRUCTURAL MODELS
TEST FACILITIES
TEST REACTORS